JPS6166186A - Spacer lattice for reactor fuel aggregate - Google Patents
Spacer lattice for reactor fuel aggregateInfo
- Publication number
- JPS6166186A JPS6166186A JP59187958A JP18795884A JPS6166186A JP S6166186 A JPS6166186 A JP S6166186A JP 59187958 A JP59187958 A JP 59187958A JP 18795884 A JP18795884 A JP 18795884A JP S6166186 A JPS6166186 A JP S6166186A
- Authority
- JP
- Japan
- Prior art keywords
- leaf spring
- spacer
- fuel
- nuclear reactor
- lattice
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
(57)【要約】本公報は電子出願前の出願データであるた
め要約のデータは記録されません。(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.
Description
【発明の詳細な説明】
〔発明の利用分野〕
本発明は、原子炉燃料集合体用スペーサ格子の改良に関
するものである。DETAILED DESCRIPTION OF THE INVENTION [Field of Application of the Invention] The present invention relates to improvements in spacer grids for nuclear reactor fuel assemblies.
軽水炉燃料集合体には、複数本からなる燃料棒をその長
手方向に平行に支持し、かつ各燃料棒間の間隔を一定に
保持して冷却材流路を確保するためスペーサ格子が装着
されておシ、スペーサ格子が燃料棒と直接接触する部分
には、板ばね構造が採用されている。A light water reactor fuel assembly is equipped with a spacer grid to support a plurality of fuel rods in parallel in the longitudinal direction and maintain a constant spacing between each fuel rod to ensure a coolant flow path. A leaf spring structure is used in the part where the spacer grid directly contacts the fuel rods.
スペーサ格子の材料は、初期の燃料集合体にあっては、
ステンレス鋼が使用されていたが、現在の沸騰水型原子
炉燃料集合体にあっては、中性子経済を考慮して、熱中
性子吸収断面積の小さいジルコニウムを基合金とするジ
ルカロイが使用されており、板ばねの部分のみ、硬度の
高いインコネルを使用するようにしている。また、従来
の加圧水型原子炉燃料集合体にあっては、スペーサ格子
にインコネルを使用していたが、最近では、沸騰水型原
子炉燃料集合体と同様中性子経済を考慮して、ジルカロ
イを材料とするスペーサ格子も使用されている。The material of the spacer grid was
Previously, stainless steel was used, but in consideration of neutron economy, Zircaloy, a zirconium-based alloy with a small thermal neutron absorption cross section, is used in current boiling water reactor fuel assemblies. Inconel, which has high hardness, is used only for the leaf spring part. In addition, conventional pressurized water reactor fuel assemblies used Inconel for the spacer lattice, but recently Zircaloy has been used as a material in consideration of neutron economy, similar to boiling water reactor fuel assemblies. Spacer grids have also been used.
ところで、原子炉運転中、スペーサ格子の板ばねは、２
８０Ｃ以上の温度にまで昇温し高速中性子の照射環境下
で負荷をかけられている。したがって、上記した板ばね
は、熱クリープおよび中性子照射クリープによって変形
し、ばね効果が低下するいわゆる緩和現象を呈するため
、燃料棒を支持する力が低下す、２Ｉ。その実例を以下
に示す。By the way, during reactor operation, the leaf springs of the spacer grid are
The temperature is raised to over 80C and a load is applied under an environment of fast neutron irradiation. Therefore, the above-mentioned leaf spring is deformed by thermal creep and neutron irradiation creep, and exhibits a so-called relaxation phenomenon in which the spring effect is reduced, resulting in a reduction in the force supporting the fuel rods, 2I. An example is shown below.
第４図は商用軽水炉燃料集合体における燃料燃焼度−燃
料棒引抜力特性線図で、第４図の測定に用いられた燃料
集合体用スペーサ格子の板ばね材料はインコネルであっ
た。また、燃料棒の引抜力は、スペーサ格子の板ばねと
燃料棒被覆管との間の摩擦力を測定したもので、燃料棒
被覆管の表面が平滑であると仮定すると、燃料棒の引抜
力は、スペーサ格子の板ばねと燃料棒被覆管との接触力
に比例する。FIG. 4 is a fuel burnup-fuel rod withdrawal force characteristic diagram for a commercial light water reactor fuel assembly, and the leaf spring material of the spacer grid for the fuel assembly used for the measurements in FIG. 4 was Inconel. In addition, the fuel rod pullout force is a measurement of the frictional force between the leaf spring of the spacer lattice and the fuel rod cladding tube.Assuming that the surface of the fuel rod cladding tube is smooth, the fuel rod pullout force is proportional to the contact force between the spacer grid leaf spring and the fuel rod cladding.
第４図から明らかなように、燃料燃焼度が進むにつれて
、燃料棒の引抜力は低下し、約３０Ｑｗｄ／ｌで中性子
照射前の約３０係程度にまで低下している。As is clear from FIG. 4, as the fuel burn-up progresses, the pull-out force of the fuel rod decreases, and at about 30 Qwd/l, it has decreased to about 30 factors before neutron irradiation.
第５図は第４図に示した商用軽水炉燃料集合体における
高速中性子照射線量−スペーサ格子板ばね緩和率特性線
図で、板ばねの緩和率は、平均で約８０優にも達してい
ることが判る。Figure 5 is a fast neutron irradiation dose vs. spacer lattice plate spring relaxation rate characteristic diagram for the commercial light water reactor fuel assembly shown in Figure 4, showing that the relaxation rate of the plate springs reaches about 80 on average. I understand.
しかして、原子炉運転中、上記のごとく、スペーサ格子
の板ばねが緩和すると、板ばねと燃料棒との間の接触力
が極端に小さくなり、このように、板ばねと燃料棒との
間の接触力が低下すると、燃料棒の支持が不安定となシ
、熱流力振動などによってフレッティング腐食が発生し
、燃料棒が破損する危険性がある。During reactor operation, when the leaf springs of the spacer lattice relax as described above, the contact force between the leaf springs and the fuel rods becomes extremely small. If the contact force of the fuel rod decreases, the support of the fuel rod becomes unstable, and fretting corrosion occurs due to heat-hydraulic vibration, leading to the risk of damage to the fuel rod.
なお、本件に関する参考文献として、
α） 、Ｉｉ’、 Ｇａｒｚａｒｏｌｌｉ ｅｔａｌ
、、 ”ＫＷＵ Ｆｕｅｌｐｅｒｆｏｒｍａｎｃｅ
ｗｉｔｈ ｐ）ｎｐｈａｓｉｓ ｏｎ ｌ（ｉｇｈＢｕ
ｒｎｕｐ”、ＡＮＳ Ｔｏｐｉｃａｌ Ｍｅｅｔｉｎｇ
Ｐｏｒｔｌａｎｄ。In addition, as references regarding this matter, α), Ii', Garzarolli et al.
,, ”KWU Fuel performance
with p)nphasis on l(ighBu
ANS Topical Meeting
Portland.
（２） ）ＬＷＷｉｌｓｏｎ ｅｔａｌ、−ｐｕｅｌ
ｐｅｒｆｏｒｍａｎｃｅＣｈａｒａｃｔｅｒｉｓｔ
ｉｃｓ ａｔ Ｅｘｔｅｎｄｅｄ Ｂｕｒｎｕｐ＝。(2) ) LW Wilson etal, -puel
performanceCharacterist
ics at Extended Burnup=.
ＡＮＳ Ｔｏｐｉｃａｌ Ｍｅｅｔｉｎｇ 、Ｗｉｌｌ
ｉａｍｓｂｕｒｇ。ANS Topical Meeting, Will
iamsburg.
（３） Ｒ，Ａ、Ｍｕｒｇａｔｒｏｙｄ ａｎｄ Ａ
、Ｒｏｇｅｒｓｏｎ 。(3) R, A, Murgatroyd and A
, Rogerson.
Ｊ、Ｎｕｃｌ 、Ｍａｔｅｒ、９０ （１９８０）２
４０−２４８などがある。J, Nucl, Mater, 90 (1980) 2
40-248 etc.
本発明は、上記した従来技術の問題点を解決すへく、種
々研究・開発の結果なされたものであって、その目的と
するところは、原子炉運転時における高速中性子照射に
よるスペーサ格子板ばねの緩和を補い、板ばねと燃料棒
との間の接触を常に維持して、フレッティング腐食の発
生を防止し、ひいては燃料棒の破損を未然に防止するこ
とのできる、安全性にすぐれた原子炉燃料集合体用スペ
ーサ格子を提供しようとするものである。The present invention has been made as a result of various research and development in order to solve the problems of the prior art described above, and its purpose is to solve the problems of the prior art described above. A highly safe atom that can compensate for the relaxation of air pressure, maintain contact between the leaf spring and the fuel rod at all times, prevent fretting corrosion, and even prevent fuel rod breakage. It is an object of the present invention to provide a spacer grid for a reactor fuel assembly.
上記目的を達成するため、本発明は、複数本からなる燃
料棒をその長手方向に平行に支持し、かつ各燃料棒間の
間隔を一定に保持して冷却材流路を確保する構造の原子
炉燃料集合体用スペーサ格子において、上記燃料棒を接
触支持する板ばねを、高速中性子照射によって伸縮を生
じる異方性を有する材料で構成し、かつ高速中性子照射
によって板ばね材料の伸びる方向を、当該板ばねの長手
方向に配置してなることを特徴とするものである。In order to achieve the above object, the present invention provides an atom structure that supports a plurality of fuel rods in parallel in their longitudinal direction and maintains a constant interval between each fuel rod to ensure a coolant flow path. In the spacer lattice for reactor fuel assemblies, the leaf springs that contact and support the fuel rods are made of an anisotropic material that expands and contracts when irradiated with fast neutrons, and the direction in which the leaf spring material stretches due to the irradiation with fast neutrons is It is characterized by being arranged in the longitudinal direction of the leaf spring.
以下、本発明を、第６図および第７図を参照しつつ、第
１図の一実施例にもとづいて説明する。The present invention will be described below based on the embodiment shown in FIG. 1, with reference to FIGS. 6 and 7.
ところで、ジルコニウムのごとき六方晶系の金属は、高
速中性子照射によって特定の方向に照射成長する。その
−例を第６図に示す。By the way, a hexagonal metal such as zirconium grows in a specific direction by fast neutron irradiation. An example is shown in FIG.
第６図は冷間加工されたジルカロイ・２管材を３５３に
と５５３にとで中性子照射したときの高速中性子照射線
量−ジルカロイ・２管材中性子照射成長特性線図、第７
図はジルコニウム六方晶の結晶模式図で、第６図の測定
に使用されたジルカロイ・２管材の集合組織は管の軸方
向に六方晶のｃ＃＋（第７図参照）の１２チが向いてお
シ、周方向には５２ｇ！Ｉ、径方向には３６チが向いて
いる。Figure 6 is a fast neutron irradiation dose-Zircaloy 2 tube material neutron irradiation growth characteristic diagram when cold worked Zircaloy 2 tube material is irradiated with neutrons at 353 and 553.
The figure is a schematic diagram of a zirconium hexagonal crystal. The texture of the Zircaloy 2 tube material used for the measurements in Figure 6 is such that the 12th crystal of the hexagonal c#+ (see Figure 7) is oriented in the axial direction of the tube. 52g in the circumferential direction! I, 36 inches are facing in the radial direction.
第６図から、ジルカロイ・２管材は、高速中性子照射に
よって軸方向に伸び、周方向に縮んでいることが判る。From FIG. 6, it can be seen that the Zircaloy 2 tube material is elongated in the axial direction and contracted in the circumferential direction due to fast neutron irradiation.
ジルカロイ・２管材の中性子照射成長は、中性子照射に
よって生じた格子欠陥が蓄積することにより、第７図に
示すように、ジルコニウム六方晶の結晶がＣ軸方向に縮
み、ａ軸方向に拡がることを意味する。During neutron irradiation growth of Zircaloy 2 tube material, lattice defects caused by neutron irradiation accumulate, and as shown in Figure 7, the zirconium hexagonal crystal shrinks in the C-axis direction and expands in the a-axis direction. means.
したがって、第１図に示すように、スペーサ格子の板ば
ね１、燃料棒被覆管２、燃料ぺＶット３からなる燃料集
合体において 板ばね１の材料をジルカロイとし、その
集合組織を図中に示した配向とし、中性子照射成長によ
る板ばね１の長手方向の伸びを拘束すれば、板ばね１に
燃料棒被覆管ス“ノ′
２側への曲シを生じ、当該板ばね１には、原子炉運転時
におけるクリープ変形を補う力が作用する。Therefore, as shown in Fig. 1, in a fuel assembly consisting of a leaf spring 1 of a spacer lattice, a fuel rod cladding tube 2, and a fuel plug 3, the material of the leaf spring 1 is Zircaloy, and its texture is shown in the figure. If the leaf spring 1 is oriented as shown in FIG. 1 and restrained from elongating in the longitudinal direction due to neutron irradiation growth, the leaf spring 1 will bend toward the fuel rod cladding groove 2, and the leaf spring 1 will be , a force acts to compensate for creep deformation during reactor operation.
なお、上記実施例に示すごとく、スペーサ格子の板ばね
１を燃料棒被覆管２と同じジルコニウム基合金で構成す
れば、板ばねｌによる熱中性子吸収断面積は小さく、中
性子経済性にすぐれているばかシでなく、炉水中におけ
る板ばね１と燃料棒被覆管２との電気化学的寿腐食反応
をも防止することができる。さらに、従来、スペーサ格
子の板ばね材料に用いられているインコネルは、ジルコ
ニウム基合金よυも硬度の高い合金であるため、インコ
ネルからなる板ばねとジルコニウム基合金からなる燃料
棒被覆管とが接触すると、燃料棒被覆管側に摩耗を生じ
るが、上記のごとく、スペーサ格子の板ばね１を燃料棒
被覆管２と同じジルコニウム基合金で構成すれば、板ば
ね１と燃料棒被覆管２との接触による摩耗を従来よシも
低減することができる。As shown in the above embodiment, if the plate spring 1 of the spacer lattice is made of the same zirconium-based alloy as the fuel rod cladding tube 2, the thermal neutron absorption cross section by the plate spring 1 is small, and neutron economy is excellent. It is possible to prevent the electrochemical corrosion reaction between the leaf spring 1 and the fuel rod cladding tube 2 in the reactor water. Furthermore, since Inconel, which has conventionally been used as a leaf spring material for spacer grids, is an alloy with a higher hardness than zirconium-based alloys, the leaf springs made of Inconel come into contact with the fuel rod cladding tubes made of zirconium-based alloys. This causes wear on the fuel rod cladding tube side, but if the leaf spring 1 of the spacer lattice is made of the same zirconium-based alloy as the fuel rod cladding tube 2 as described above, the contact between the leaf spring 1 and the fuel rod cladding tube 2 will be reduced. Wear caused by contact can be reduced compared to conventional methods.
本発明の他の実施例を第２図および第３図に示二゛ す。Other embodiments of the invention are shown in FIGS. 2 and 3. vinegar.
接合したものであって、この実施例によれば、第１図の
実施例に比較して、板ばね１の強度をさらに向上させる
ことができるっ
また、第３図の実施例においては、燃料棒を接触支持す
るジルカロイ板ばね１の背面に、高速中性子照射によっ
て結晶が板ばね１の長手方向に縮む材料５を接合した場
合を示し、この実施例によれば、第１図の実施例に比較
して、原子炉運転時におけるクリープ変形を補う力をさ
らに大きくすることができる。According to this embodiment, the strength of the leaf spring 1 can be further improved compared to the embodiment shown in FIG. A case is shown in which a material 5 whose crystals shrink in the longitudinal direction of the leaf spring 1 by fast neutron irradiation is bonded to the back surface of the Zircaloy leaf spring 1 that supports the rod in contact with the rod. In comparison, it is possible to further increase the force that compensates for creep deformation during nuclear reactor operation.
以上詳述したように、本発明によれば、原子炉運転時に
おける高速中性子照射によるスペーサ格子板ばねの緩和
を補い、板ばねと燃料棒との間の接触を常に維持して、
フＶツテイング腐食の発生を防止し、ひいては燃料棒の
破損を未然に防止することのでへる、安全性にすぐれた
原子炉燃料集合体用スペーサ格子を得ることができる。As detailed above, according to the present invention, the relaxation of the spacer lattice leaf spring due to fast neutron irradiation during nuclear reactor operation is compensated for, and the contact between the leaf spring and the fuel rod is always maintained.
It is possible to obtain a spacer grid for a nuclear reactor fuel assembly that is highly safe because it prevents the occurrence of V-steing corrosion and, in turn, prevents damage to fuel rods.
第１図は本発明の一実施例である原子炉燃料集合体用ス
ペーサ格子と燃料棒との取合状態を示す縦断面図、第２
図は本発明スペーサ格子の他の実施例を示す縦断面図、
第３図は本発明スペーサ格子のさらに他の実施例を示す
縦断面図、第４図は燃料燃焼度−燃料棒引抜力特性線図
、第５図は高速中性子照射線量−スペーサ格子板ばね緩
和率特性線図、第６図は高速中性子照射線量−ジルカロ
イ・２管材中性子照射成長特性線図、第７図はジルコニ
ウム大方晶の結晶模式図である。
１・・・板ばね、２・・・燃料棒被覆管、３・・・燃料
ペレット、４・・・インコネル、５・・・結晶が板はね
１の長手方向に縮む材料。FIG. 1 is a vertical cross-sectional view showing how a spacer grid for a nuclear reactor fuel assembly and fuel rods are connected, which is an embodiment of the present invention, and FIG.
The figure is a vertical sectional view showing another embodiment of the spacer grid of the present invention.
Fig. 3 is a vertical cross-sectional view showing still another embodiment of the spacer lattice of the present invention, Fig. 4 is a fuel burnup-fuel rod pull-out force characteristic diagram, and Fig. 5 is a fast neutron irradiation dose - spacer lattice plate spring relaxation. FIG. 6 is a fast neutron irradiation dose-neutron irradiation growth characteristic diagram for Zircaloy two-tube material, and FIG. 7 is a schematic diagram of zirconium macrogonal crystal. DESCRIPTION OF SYMBOLS 1... Leaf spring, 2... Fuel rod cladding tube, 3... Fuel pellet, 4... Inconel, 5... Material whose crystals shrink in the longitudinal direction of the plate spring 1.
Claims (1)
し、かつ各燃料棒間の間隔を一定に保持して冷却材流路
を確保する構造の原子炉燃料集合体用スペーサ格子にお
いて、上記燃料棒を接触支持する板ばねを、高速中性子
照射によつて伸縮を生じる異方性を有する材料で構成し
、かつ高速中性子照射によつて板ばね材料の伸びる方向
を、当該板ばねの長手方向に配置してなることを特徴と
する原子炉燃料集合体用スペーサ格子。 ２、特許請求の範囲第１項記載の発明において、燃料棒
を接触支持する板ばねをジルコニウム基合金で構成した
原子炉燃料集合体用スペーサ格子。 ３、特許請求の範囲第１項または第２項に記載の発明に
おいて、燃料棒を接触支持する板ばねの背面にインコネ
ルを接合した原子炉燃料集合体用スペーサ格子。 ４、特許請求の範囲第１項または第２項に記載の発明に
おいて、燃料棒を接触支持する板ばねの背面に、高速中
性子照射によつて結晶が板ばねの長手方向に縮む材料を
接合した原子炉燃料集合体用スペーサ格子。[Claims] 1. A nuclear reactor fuel assembly having a structure in which a plurality of fuel rods are supported in parallel in the longitudinal direction and the intervals between each fuel rod are maintained constant to ensure a coolant flow path. In the body spacer lattice, the leaf springs that contact and support the fuel rods are made of an anisotropic material that expands and contracts when irradiated with fast neutrons, and the direction in which the leaf spring material stretches is controlled by the irradiation with fast neutrons. A spacer lattice for a nuclear reactor fuel assembly, characterized in that the spacer lattice is arranged in the longitudinal direction of the leaf spring. 2. A spacer lattice for a nuclear reactor fuel assembly according to the invention as set forth in claim 1, in which the leaf springs that contact and support the fuel rods are made of a zirconium-based alloy. 3. In the invention as set forth in claim 1 or 2, a spacer lattice for a nuclear reactor fuel assembly in which Inconel is bonded to the back surface of a leaf spring that contacts and supports fuel rods. 4. In the invention set forth in claim 1 or 2, a material whose crystals shrink in the longitudinal direction of the leaf spring by high-speed neutron irradiation is bonded to the back surface of the leaf spring that contacts and supports the fuel rods. Spacer grid for nuclear reactor fuel assembly.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP59187958A JPS6166186A (en) | 1984-09-10 | 1984-09-10 | Spacer lattice for reactor fuel aggregate |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP59187958A JPS6166186A (en) | 1984-09-10 | 1984-09-10 | Spacer lattice for reactor fuel aggregate |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| JPS6166186A true JPS6166186A (en) | 1986-04-04 |
Family
ID=16215139
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP59187958A Pending JPS6166186A (en) | 1984-09-10 | 1984-09-10 | Spacer lattice for reactor fuel aggregate |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS6166186A (en) |
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4465869A (en) * | 1980-03-21 | 1984-08-14 | Kao Corporation | Process for the production of glyceryl ethers |
| JPH07185487A (en) * | 1993-12-27 | 1995-07-25 | N K Nada Seika Kk | Device for removing residual matter in container bag |
| JPH0745081U (en) * | 1990-12-26 | 1995-12-19 | 株式会社シオヤ | Dust removal device for carrying bags for cement-containing coagulant |
| JPH081123A (en) * | 1994-06-21 | 1996-01-09 | Asano Seiki Kk | Wrapping bag cleaning device and its wrapping bag transport mechanism |
-
1984
- 1984-09-10 JP JP59187958A patent/JPS6166186A/en active Pending
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4465869A (en) * | 1980-03-21 | 1984-08-14 | Kao Corporation | Process for the production of glyceryl ethers |
| JPH0745081U (en) * | 1990-12-26 | 1995-12-19 | 株式会社シオヤ | Dust removal device for carrying bags for cement-containing coagulant |
| JPH07185487A (en) * | 1993-12-27 | 1995-07-25 | N K Nada Seika Kk | Device for removing residual matter in container bag |
| JPH081123A (en) * | 1994-06-21 | 1996-01-09 | Asano Seiki Kk | Wrapping bag cleaning device and its wrapping bag transport mechanism |
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